ML031490485
| ML031490485 | |
| Person / Time | |
|---|---|
| Site: | Neely Research Reactor |
| Issue date: | 06/24/2003 |
| From: | Madden P NRC/NRR/DRIP/RORP |
| To: | Hertel N Neely Research Reactor |
| Holmes S, NRC/NRR/DRIP/RORP, 415-8583 | |
| References | |
| IR-02-201 | |
| Download: ML031490485 (30) | |
See also: IR 05000160/2002201
Text
June 24, 2003
Dr. Nolan Hertel, Director
Neely Nuclear Research Center
Georgia Institute of Technology
900 Atlantic Drive
Atlanta, GA 30332-0425
SUBJECT: NRC INSPECTION REPORT NO. 50-160/2002-201
Dear Dr. Hertel:
The inspection effort involved the coordination of the confirmatory radiological survey activities
performed by our contractor, Oak Ridge Institute for Science and Education, of your research
reactor on October 21-23, 2002. In addition, various aspects of your reactor operations,
decommissioning, and radiation protection programs were inspected, including selective
examinations of procedures and representative records, interviews with personnel, and
observations of the facility.
Based on the results of this inspection, it has been determined that: 1) the decommissioning of
the 5 MWt Research Reactor has been performed in accordance with the approved
Decommissioning Plan; 2) the terminal radiation survey and associated documentation from the
licensee demonstrated that residual radioactive material at the facility and site is less than the
NRC-approved guideline limits; and 3) since the licensee has met their NRC-approved guideline
limits, the facility and site meet the criteria for license termination set forth in 10 CFR Part 20.1401(b)(2).
No safety concern or noncompliance with Nuclear Regulatory Commission (NRC) requirements
was identified. No response to this letter is required.
Dr. N. Hertel -2-
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRCs document system
(ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading
Room) http://www.nrc.gov/NRC/ADAMS/index.html. Should you have any questions
concerning this inspection, please contact Mr. Stephen Holmes at 301-415-8583.
Sincerely,
/RA by Daniel E. Hughes, Acting for/
Patrick M. Madden, Section Chief
Research and Test Reactors Section
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Docket No. 50-160
License No. R-97
Enclosures: 1. NRC Inspection Report No. 50-160/2002-201
2. Confirmatory Survey Plan for the Georgia Tech Research Reactor dated
October 9, 2002
3. Confirmatory Survey of the Georgia Tech Research Reactor, dated
February 2003
cc w/enclosures: Please see next page
Georgia Institute of Technology Docket No. 50-160
cc:
Mr. Charles H. Badger Ms. Glen Carrol
Office of Planning and Budget 139 Kings Highway
Room 608 Decatur, GA 30030
270 Washington Street, S.W.
Atlanta, GA 30334 Charles Bechhoefer, Chairman
Atomic Safety and
Mayor of City of Atlanta Licensing Board Panel
55 Trinity Avenue, S.W. U.S. NRC, MS: T3-F23
Suite 2400 Washington, DC 20555-0001
Atlanta, GA 30335
Mr. James C. Hardeman, Jr.
Dr. William Vernetson Manager, Environmental
Director of Nuclear Facilities Radiation Program
Department of Nuclear Engineering Environmental Protection Division
Sciences Dept. of Natural Resources
University of Florida State of Georgia
202 Nuclear Sciences Center 4244 International Parkway
Gainesville, FL 32611 Suite 114
Atlanta, GA 30354
Joe D. Tanner, Commissioner
Department of Natural Resources Dr. Jean-Lou Chameau, Dean
47 Trinity Avenue, S.W. College of Engineering
Atlanta, GA 30334 Georgia Institute of Technology
225 North Avenue
Dr. Rodney Ice, MORS Atlanta, GA 30332-0425
Neely Nuclear Research Center
Georgia Institute of Technology Dr. Peter S. Lam
900 Atlantic Drive Atomic Safety and Licensing Board Panel
Atlanta, GA 30332-0425 U.S. NRC, MS: T3-F23
Washington, DC 20555-0001
Ms. Pamela Blockey-OBrien
D23 Golden Valley Dr. J. Narl Davidson, Interim Dean
Douglasville, GA 30134 Chair, Technical and Safety Review
Committee
Mr. E.F. Cobb Georgia Institute of Technology
Southern Nuclear Company 225 North Avenue
42 Iverness Center Atlanta, GA 3033-0360
Birmingham, AL 35242
Dr. Charles Liotta, Vice Provost
Dr. G. Wayne Clough, President of Research and Dean of
Georgia Institute of Technology Graduate Studies
Carnegie Building Georgia Institute of Technology
Atlanta, GA 30332-0325 225 North Avenue
Atlanta, GA 30332
Dr. N. Hertel -2- June 24, 2003
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRCs document system
(ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading
Room) http://www.nrc.gov/NRC/ADAMS/index.html. Should you have any questions
concerning this inspection, please contact Mr. Stephen Holmes at 301-415-8583.
Sincerely,
/RA by Daniel E. Hughes, Acting for/
Patrick M. Madden, Section Chief
Research and Test Reactors Section
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Docket No. 50-160
License No. R-97
Enclosures: 1. NRC Inspection Report No. 50-160/2002-201
2. Confirmatory Survey Plan for the Georgia Tech Research Reactor dated
October 9, 2002
3. Confirmatory Survey of the Georgia Tech Research Reactor, dated
February 2003
cc w/enclosures: Please see next page
DISTRIBUTION:
PUBLIC RORP/R&TR r/f TDragoun PDoyle WEresian PIsaac
SHolmes CBassett MMendonca FGillespie WBeckner EHylton
AAdams BDavis (Ltr.only O5-A4)
ACCESSION NO.: ML031490485 TEMPLATE #: NRR-106
OFFICE RORP:LA RORP:RI RORP:SC
NAME EHylton:rdr SHolmes PMadden
DATE 06/ 04 /2003 06/ 04 /2003 06/ 24 /2003
C = COVER E = COVER & ENCLOSURE N = NO COPY
OFFICIAL RECORD COPY
U. S. NUCLEAR REGULATORY COMMISSION
Docket No: 50-160
License No: R-97
Report No: 50-160/2002-201
Licensee: Georgia Institute of Technology
Facility: Georgia Institute of Technology Research Reactor (GTRR)
Location: 900 Atlantic Drive
Atlanta, GA 30332
Dates: October 21-23, 2002
Inspector: Stephen W. Holmes
Approved by: Patrick M. Madden, Section Chief
Research and Test Reactors Section
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
EXECUTIVE SUMMARY
Georgia Institute of Technology Research Reactor
Report No: 50-160/2002-201
This routine, announced inspection involved the confirmatory radiological survey and the on-site
review of selected activities being performed at the Georgia Institute of Technology Research
Reactor. In addition, the activities audited during this inspection included: organization and
staffing; review and audit functions; procedures; removal of materials; decommissioning
activities; release criteria; confirmatory final survey; maintenance and surveillance; and
radiation protection program. The inspector was assisted by the NRCs contractor, Oak Ridge
Institute for Science and Education Environmental Survey and Site Assessment Program.
Organization and Staffing
! The organizational structure and their corresponding functions were consistent with
Technical Specification Section 5.0, Amendment No. 14, dated July 22, 1999, and the
Decommissioning Plan for the Georgia Institute of Technology Research Reactor facility
dated June 1998.
Review and Audit Functions
! The audits conducted by the Technical Safety Review Committee and Georgia Institute
of Technology Research Reactor staff were in accordance with the requirements
specified in Technical Specification Section 5.2. and Decommissioning Plan Section 2.4.
Procedures
! The procedural control and implementation program was acceptably maintained and
met Technical Specifications and Decommissioning Plan requirements.
Removal of Materials
! Fuel and radioactive and non-radioactive waste was removed from the site in
accordance with the Georgia Institute of Technology Research Reactor
Decommissioning Plan requirements, and Department of Transportation and Nuclear
Regulatory Commission regulations.
Decommissioning Activities
! Decommissioning activities were performed as required by Decommissioning Plan
Section 2.3 and licensee procedures.
Release Criteria
! Duratec used the appropriate guideline and screening values, as required by the
NRC-approved Decommissioning Plan, in performing the final survey.
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Confirmatory Final Survey
! The elevated surface activity and exposure readings in the basement compressor room
were due to naturally occurring radioactive material.
! Based on the results of the licensees final status survey and Nuclear Regulatory
Commissions confirmatory measurements, Georgia Institute of Technology has
adequately demonstrated that the Georgia Institute of Technology Research Reactor
facility satisfies the criteria for release for unrestricted use.
Maintenance and Surveillance
! The maintenance program was implemented as required by Georgia Institute of
Technology procedures.
! The licensee's program for surveillance and limiting conditions for operation
confirmations satisfied Technical Specification and Decommissioning Plan
requirements.
! The licensee's design change procedures were in place and were implemented as
required by licensee procedures.
Radiation Protection Program
! The radiation protection program satisfied the requirements of 10 CFR 19.12 and
.
! Radiological postings satisfied regulatory requirements.
! Surveys were performed and documented as required by 10 CFR 20.1501(a), Technical
Specifications, and licensee procedures.
! The personnel dosimetry program was acceptably implemented and doses were in
conformance with licensee and 10 CFR Part 20 limits.
! Portable survey meters, radiation monitoring, and counting lab instruments were
maintained according to Technical Specifications, industry/equipment manufacturer
standards, and licensee and contractor procedures.
! The evaluation and administration of the respiratory program were adequately
performed according to Decommissioning Plan and Nuclear Regulatory Commission
requirements.
! The program for monitoring, storage, and release of effluents was acceptable.
Report Details
Summary of Plant Status
Georgia Institute of Technology (GIT), in Atlanta Georgia, has completed decommissioning its
5 MWt Research Reactor (GTRR) and associated systems. The reactor was located within the
Neely Nuclear Research Center (NNRC) on GITs main campus. The reactor was designed for
several different research applications including experiments using high intensity neutron
beams, gamma ray beams, and an uniform thermal neutron flux through a large sized beam.
Although it was originally designed for 1 MWt output, it was upgraded to produce 5 MWt in
1974. The GTRR was built in the early 1960's as a research and training reactor. Operating
under the Nuclear Regulatory Commission (NRC) License No. R-97, it went critical for the first
time on December 31, 1964.
On November 17, 1995, all operations at the reactor ceased. GIT contracted NES, Inc. to
perform the initial characterization survey and to provide a decommissioning plan for the GTRR.
In October 1997, NES performed a characterization survey of the GTRR, based upon the GIT
Decommissioning Project - Radiological Characterization Plan. Results of the characterization
survey were provided in NES Georgia Institute of Technology Research Reactor
Decommissioning Project Characterization Report issued May 1998. GIT requested the NRC,
by letters dated July 1, 1998, February 8, 1999, and May 28, 1999, to grant them the
authorization to decommission the reactor according to their submitted decommissioning plan.
On July 22, 1999, the NRC issued Amendment No. 14 to the reactor licence that approved
GITs Decommissioning Plan. GIT contracted with IT Corporation (IT) to decommission the
GTRR facility. IT, through its subcontractor GTS Duratec (Duratec), started decommissioning
operations December 1999. Final waste shipment was made August 2001.
The Final Status Survey Report for the GTRR facility was completed and issued June 2002.
According to the report, all contaminated systems and components had been removed from the
site. Potentially contaminated structural surfaces identified during characterization surveys had
been removed and/or remediated such that the residual radioactivity is less than NRC
Regulatory Guide 1.86 limits.
The NRC requested Oak Ridge Institute for Science and Educations (ORISE) Environmental
Survey and Site Assessment Program (ESSAP) to perform a confirmatory survey of the GTRR
facility. On October 21-23, 2002, the ESSAP team, accompanied by an NRC inspector,
conducted this survey.
1. ORGANIZATIONAL STRUCTURE AND FUNCTIONS
a. Inspection Scope (Inspection Procedures (IP) 69001 and 40755)
The inspector reviewed selected aspects of:
- organization and staffing
- qualifications
- management responsibilities
- administrative controls
- decommissioning activity records
- GTRR Decommissioning Plan (DP) dated June 1998
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- Technical Specifications (TS), Amendment No. 14, dated July 22, 1999
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b. Observations and Findings
The general organizational structure and staffing had not changed since the last
inspection. The organizational structure and staffing at the facility were as reported in
the Annual Report and as required by TS Section 5.1 and Figure 5.1. Review of
records verified that management responsibilities were administered as required by
TS Sections 5.2 thru 5.6 and applicable procedures.
The decommissioning of the reactor required GTRR management to assume
additional project management responsibilities. Through record reviews and
interviews with the reactor manager, radiation safety officer (RSO), and Duratec
project manager, the inspector confirmed that both GTRR management and the
decommissioning project organization structures were as required by DP Section 2.4
and Figure 2.2.
c. Conclusions
The organizational staff and their corresponding functions and responsibilities were
consistent with TS Section 5.0, Amendment No. 14, dated July 22, 1999, and the DP
for the GTRR facility dated June 1998
2. REVIEW AND AUDIT FUNCTIONS
a. Inspection Scope (IPs 69001 and 40755)
The inspector reviewed selected aspects of:
- Technical Safety Review Committee (TSRC) meeting minutes
- GTRR staff safety review records
- TSRC and GTRR staff audit records
- responses to safety reviews and audits
- personnel qualifications
- GTRR DP dated June 1998
- TS, Amendment No. 14, dated July 22, 1999
b. Observations and Findings
DP Section 2.4 states that the TSRC: 1) will review and approve all plans, policies and
procedures to be performed under the GTRR Decommissioning Project, 2) will review
and audit the decontamination and decommissioning project operations and activities,
3) members will be appointed by the President of Georgia Tech, and 4) will keep a
written record of the meetings and will report directly to the President.
During inspections in 2000 and 2002, the inspector reviewed the qualifications of the
TSRC members and confirmed that they met the requirements specified in TS Section 5.2 and DP Section 2.4. The results of the 2000 inspections were
documented in NRC Inspection Report (IR) No. 50-160/2000-201 dated March 15,
2000, NRC IR No. 50-160/2000-202 dated August 31, 2000, and NRC IR
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No. 50-160/2000-203 dated December 1, 2000. The inspector noted that the TSRC
met more often than the required semiannual frequency and that a quorum was
present each time. The inspector reviewed the minutes of the TSRC and determined
that they provided guidance, direction, operations oversight, and 10 CFR 50.59
request reviews as required by the DP and TS.
TSRC meeting minutes and audit records and GTRR staff audit records showed that
safety reviews and audits were conducted as required by TS Section 5.2(d). The
content of the audits and safety reviews were consistent with the TS. These reviews
provided appropriate guidance, direction, and oversight to ensure satisfactory
decommissioning of the reactor.
By examining the TSRCs review of the DP and their audits of the operations and
training programs, the inspector determined that the safety reviews, audits, and
associated findings were satisfactory and that the licensee took the appropriate
corrective actions in response to the findings.
The inspector reviewed selected decommissioning and facility change approvals.
Records and observations showed that changes at the facility were acceptably
reviewed in accordance with 10 CFR 50.59 and applicable licensee administrative
controls. None of the changes constituted an unreviewed safety question or required
a change to the TS. The inspector determined that TSRC 10 CFR 50.59 request
reviews were adequately performed.
c. Conclusions
The audits conducted by the TSRC and GTRR staffs were in accordance with the
requirements specified in TS Section 5.2 and DP Section 2.4. TSRC 10 CFR 50.59
request reviews were adequately performed.
3. PROCEDURES
a. Inspection Scope (IPs 69001 and 40755)
The inspector reviewed selected aspects of:
- administrative controls
- records for changes and temporary changes
- DP dated June 1998
- TS, Amendment No. 14, dated July 22, 1999
- decommissioning procedures
- logs and records
b. Observations and Findings
During decommissioning activities, the inspector confirmed that written health physics
(HP) and decommissioning procedures were available for those tasks and items
required by TS Section 5.3 and the DP Sections 2.3.1.1. and 3.1.2.2. The procedures
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were routinely updated and then approved by the TSRC while minor modifications to
the procedures were approved by the facility director.
Decommissioning procedures and operating plans reviewed and approved by the
TSRC included those dealing with:
- Initial Radiological Survey Plan and Procedures
- Health and Safety Plan and Procedures
- Waste Management Plan and Procedures
- Management Plan
- Quality Assurance Plan and Procedures
- Radiation Protection Plan and Procedures
- Decommissioning Work Plan
- Final Radiological Survey Plan
Through review of the 2000 training records and interviews with staff, the inspector
determined that the training of staff and contractor personnel concerning procedures
was adequate. During the inspectors tours of the facility, it was observed that
personnel performing radiation surveys, conducting instrument checks, issuing
dosimetry, and performing the decommissioning work were doing so in accordance
with applicable procedures.
c. Conclusions
Based on the procedures and records reviewed and observations of personnel during
the inspections in 2000, it was determined that the procedural control and
implementation program was acceptably maintained and met TS and DP
requirements.
4. REMOVAL OF MATERIALS
a. Inspection Scope (IPs 69001, 86740, and 85102)
The inspector reviewed selected aspects of:
- transportation records
- disposal records
- NRC Forms 741 and 742
- DP dated June 1998
b. Observations and Findings
From 1964 through 1995, the licensee operated a heavy water moderated and cooled
research reactor at the NNRC. The reactor was shut down on November 17, 1995, in
preparation for the summer Olympic Games in Atlanta, GA, and was never restarted.
As noted in a previous NRC IR No. 50-160/1996-01, the irradiated fuel was shipped to
the Savannah River Site on February 18, 1996. The licensee had previously shipped
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the unirradiated fuel to the Oak Ridge National Laboratory site in Tennessee on
January 31, 1996. The inspector confirmed that, as noted by DP Section 1.5, all fuel
had been removed from NNRC prior to decommissioning.
Fifty-six (56) total radioactive waste shipments were made during the GTRR
decommissioning. The final waste shipment occurred on August 3, 2001. Radioactive
waste was sent to one of four consignees: 1 Duratek Inc.; 2 CNSI Barnwell; 3
Envirocare of Utah; and 4 Westinghouse Savannah River Site. During 2000, the
inspector confirmed through records review, interviews with licensee staff, and actual
observation, that radioactive waste was disposed of as required by DP Section 3.2
and in accordance with Department of Transportation and NRC regulations.
c. Conclusions
As a result of the records review and on-site observations made during
decommissioning tours, it was confirmed that the fuel and radioactive waste were
removed from the site in accordance with the GTRR DP requirements, and
Department of Transportation and NRC regulations.
5. DECOMMISSIONING ACTIVITIES
a. Inspection Scope (IPs 69001 and 40755)
The inspector reviewed selected aspects of:
- operational logs and records
- decommissioning procedures
- decommissioning logs and records
- DP dated June 1998
- the facility during tours
b. Observations and Findings
As noted above, the reactor was permanently shut down on November 17, 1995. All
irradiated reactor fuel was removed from the site on February 18, 1996. On July 22,
1999, following a request by the licensee and a review by the NRC, Amendment No.
14 to Facility License No. R-97 was issued which authorized decommissioning of the
GTRR. The licensees contractor started its decommissioning of the facility in January
2000. (Actual decommissioning of the facility was completed in May 2001, although
the contractors final survey of the facility continued for several months afterwards.)
Decommissioning activities focused on the dismantling and removal of the reactor
proper, its support structures, auxiliary equipment and components, and the biological
shield. The inspector examined the following selected tasks as directly described in
DP Section 2.3, Decommissioning Activities and Tasks:
Reactor Complex
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Vertical Beam Ports - The vertical beam ports will be removed - including the
thimbles, thimble plugs, sample tubes, and liners. The lead will be removed from
the plugs and sent to a mixed waste processor. The other items will be
segmented as necessary, packaged, and disposed of as radioactive waste.
Shim Safety Rods and Drives - The four shim safety rods will be disconnected
from the drives, removed through the top shield, cut in half, and disposed of as
mixed waste. The shim safety rod drives will be disconnected, removed,
segmented, and disposed of as radioactive waste.
Horizontal Beam Gates - The ten horizontal beam gate drive motors will be
disconnected and removed. The gates will be separated from the shafts and cut
open. The lead inside will be removed and disposed of as mixed waste, and the
remainder disposed of as radioactive waste.
Spent Fuel Storage Holes - The spent fuel storage hole plugs will be removed and
disposed of as radioactive waste. The hole liners will be core drilled out and each
liner will be cut in half, packaged, and disposed of as radioactive waste.
Piping and Instrumentation - This task involved the removal of miscellaneous
piping and ventilation in and around the reactor complex. The materials will be
disposed of as radioactive waste.
Lead Cover Plate - The lead cover plate will be removed in two distinct pieces - the
inner plate and outer plate. The 24 lead and steel port plugs will be removed from
the inner plate and cut open with an abrasive saw. The lead will be removed and
disposed of as mixed waste, and the steel will be disposed of as radioactive waste.
Upper Top Shield - The upper top shield will also be removed in two distinct pieces
- the inner shield plug and outer shield plug. The 24 concrete and steel inner port
plugs and eight concrete and steel outer port plugs will be removed and disposed
of as radioactive waste. The inner concrete and steel upper top shield will be
removed and disposed of as radioactive waste. The outer concrete and steel
upper shield plug will be removed and disposed of as radioactive waste.
Lower Shield Plug - The 31 lead, concrete, and steel port plugs will be removed
from the lower top shield plug and cut open with an abrasive saw. The lead will be
removed and disposed of as mixed waste. The remaining concrete and steel will
be disposed of as radioactive waste.
Fuel Spray Manifold - The fuel spray manifold pipe will be cut free within the
reactor, utilizing long-handled tools, and transferred to the contamination control
envelope. The manifold will be further segmented and disposed of as radioactive
waste.
Reactor Vessel - A remote operated robotic arm will be installed in the reactor
vessel to facilitate segmentation. Using an abrasive saw connected to the robotic
arm, the horizontal beam ports and through tubes will be cut free and lifted out.
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The bottom pipes will be core bored and removed. The reactor vessel will be cut
into sections using an abrasive saw mounted on the robotic arm. Lifting holes will
first be drilled into each section with a drill attached to the robotic arm, and each
section rigged. Each section will be lifted out with the overhead crane, transferred
to the packaging area and disposed of as radioactive waste.
Graphite Retaining Sleeve - The graphite retaining sleeve will be removed in a
similar fashion as the vessel. Each section will be disposed of as radioactive
waste.
Graphite Removal - The 4-inch by 4-inch graphite stringers will be removed using
long-handled tools from either the top of the biological shield or through the
thermal column. The graphite will be packaged and disposed of as radioactive
waste.
Horizontal Beam Ports - The beam port and through tube plugs will be removed
and disposed as radioactive waste. Lead will first be removed from the through
tube plugs by cutting the top off the plugs with an abrasive saw. The lead will be
disposed of as mixed waste.
Boral Removal - The 1/4-inch boral sheet staked to the inside of the steel tank will
be removed in a similar fashion as the vessel. Each section will be disposed of as
radioactive waste.
Inner Steel Tank - The inner steel tank will follow a similar removal scenario to that
described for the boral removal. The tank will be cut into sections using an
abrasive saw mounted on the robotic arm. Lifting holes will first be drilled into
each section, and each section will then be rigged. After cutting, the section will be
transferred to the packaging area using the overhead crane. Each section will be
disposed of as radioactive waste.
Lead Thermal Shield - The lead thermal shield was formed by pouring molten lead
into the space between the inner and outer steel tanks. With the inner tank and
cooling coils removed, the lead will be pried free of the outer tank in easily handled
pieces with long-handled tools. The pieces will be lowered into a basket and
transferred to a waste container. The lead will be disposed of as mixed waste.
Outer Steel Tank - The outer steel tank will be removed using the same methods
as the removal of the inner steel tank. The tank may have to be pried free of the
concrete prior to removal. Each section will be disposed of as radioactive waste.
Thermal Column Shutter and Shielding - In order to remove the thermal column
shutter and shields, the two thermal column door plugs will be removed first,
segmented with an abrasive saw and the lead removed. The steel cover plate will
then be removed, segmented and packaged. The exposed lead shield will then be
removed and packaged for processing. The concrete and steel blocks will also be
removed and packaged. Segmenting of these blocks is not required. The
concrete, steel and lead doors will be removed, segmented and packaged. Any
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remaining lead will then be removed and packaged for disposal. The concrete and
steel will be disposed of as radioactive waste and the lead as mixed waste.
Biomedical Irradiation Facility Shutter and Shielding - In order to remove the
biomedical irradiation facility shutter, the aluminum cover plate will be removed first
and segmented. The exposed lead bricks will then be removed and packaged.
The movable shield plugs and doors will also be removed. The outer bismuth
shield, the water tank, and the inner bismuth plug will be removed and packaged.
Due to the package restrictions, segmenting of these items will have to be
performed. The materials will be disposed as radioactive waste.
Fission Chambers - The fission chambers will be removed and packaged for
disposal. The remaining U-235 will be packaged and shipped to an appropriate
site.
Biological Shield
Activated Concrete - Due to the relatively small amount of activated concrete and
the limited access, the concrete will be removed with a bobcat/jackhammer. The
waste will be packaged and disposed of as radioactive waste.
Bottom Shield - As above, due to the relatively small amount of activated concrete
and the limited access the concrete will be removed with a bobcat/jackhammer.
The waste will be packaged and disposed of as radioactive waste.
During the inspections in 2000, the inspector observed various of these activities as
they were being conducted including: piping and instrumentation, upper top shield,
graphite removal, lead thermal shield, fission chambers, and activated concrete. In
order to verify that all the above tasks had been performed in accordance with the DP,
the inspector also reviewed the related licensee and contractor records and surveys,
and toured the facility. The inspector determined that the above tasks had been
completed in accordance with final approved DP.
c. Conclusions
Based on the observations made during the inspection, decommissioning activities
have been performed as required by DP Section 2.3 and licensee procedures.
6. RELEASE CRITERIA
a. Inspection Scope (IPs 69001 and 40755)
The inspector reviewed selected aspects of:
- DP dated June 1998
- Georgia Institute of Technology Research Reactor Decommissioning Project
Characterization Report, issued May 1998
- Final Status Survey Report for the GTRR facility issued June 2002
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b. Observations and Findings
The primary contaminants of concern for the GTRR are beta-gamma emittersfission
and activation productsresulting from reactor operation. The NRC-approved
guidelines for release for unrestricted use for building surfaces were based on those
for beta-gamma emitters contained in NRC Regulatory Guide 1.86 (NRC 1974). These
guidelines are:
5,000 -- dpm/100 cm2, averaged over a 1 m2 area
15,000 -- dpm/100 cm2, maximum in a 100 cm2 area
1,000 -- dpm/100 cm2, removable.
However, due to the presence of the hard-to-detect-radionuclides H-3 and Fe-55, the
above guidelines were modified to account for the contributing activity of these
radionuclides. The modified guidelines are (Shaw 2002):
2,400 -- dpm/100 cm2 average activity in a 1 m2 area
7,200 -- dpm/100 cm2 maximum activity in a 100 cm2 area
313 -- dpm/100 cm2 removable activity
GITs final survey plan (GTS 2000) stated that radionuclide concentrations in soil for
the contaminants of concern would meet the NRC published (Federal Register Vol. 64
page 68396, December 7, 1999) screening values for selected radionuclides in
surface soils. The screening values for the GTRR radionuclides of interest are
summarized below.
Radionuclide Guideline Value (pCi/g)
H-3 110
Fe-55 10,000
Pu-239/240 2.3
U-233/234 13.0
U-238 14.0
Ni-59 5,500
Cs-134 5.7
Cs-137 11.0
Co-60 3.8
Eu-152 8.7
Eu-154 8.0
Mn-54 15.0
Ag-110m 3.9
Zn-65 6.2
Sr-90 1.7
C-14 12.0
Ni-63 2,100
Tc-99 19.0
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The inspector observed and interviewed Duratec, ITs representative.
The inspector determined that Duratec used the appropriate guideline and screening
values as calculated in the Characterization Report and specified in the approved DP.
c. Conclusions
Duratec used the appropriate guideline and screening values as required by the DP, in
performing the final survey.
7. CONFIRMATORY FINAL SURVEY
a. Inspection Scope (IPs 69001 and 40755)
The inspector reviewed selected aspects of:
- DP dated June 1998
- Georgia Institute of Technology Research Reactor Decommissioning Project
Characterization Report, issued May 1998
- Final Status Survey Report for the GTRR facility issued June 2002
b. Observations and Findings
(1) Overview
DP Section 4.0, Proposed Final Radiation Survey Plan, describes the final
radiation survey to be conducted of the facility prior to license termination. This
survey is required in order to ensure that the area satisfies the unrestricted
release criteria for radioactive material according to NUREG/CR- 5849. (DP
Section 4.1) Additionally, DP Section 4.2.3 specifies, As stated in
NUREG/CR-5849, proper documentation of every aspect of the final survey is
necessary for future reference to the decommissioning survey. An accurate
mapping of the reactor containment building and surrounding areas within this
decommissioning project will be maintained for future review and verification by a
regulatory inspector.
Although the licensee is responsible for performing and documentation the
decommissioning and final status survey (Final Status Survey Report for the
GTRR facility issued June 2002), the NRC verifies the licensees performance
through inspections during decommissioning and a confirmatory final survey at the
end.
As part of this confirmatory process ESSAP reviewed and evaluated GITs final
survey plan and report (GTS 2000 and Shaw 2002). The documents were
reviewed for general thoroughness, accuracy, and consistency. Data were
evaluated to assure that areas exceeding guidelines were identified and had
undergone remediation. Final status survey results were compared with guidelines
to ensure that the data had been interpreted correctly. Comments were provided
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to the NRC, documenting the review of the final survey plan and the final survey
report.
The procedures, methods, and data submitted by GIT were considered to be
appropriate and adequately documented the radiological status of the GTRR.
ESSAP confirmed that the licensee modified the gross activity guidelines to
account for hard-to-detect radionuclides. This data was reviewed by ESSAP to
evaluate its appropriateness of use and determined it to be satisfactory.
ESSAP performed confirmatory surveys of the GTRR during the period October 21
to 23, 2002. The surveys were performed in accordance with the site-specific
survey plan submitted to and approved by the NRC and the ORISE/ESSAP Survey
Procedures and Quality Assurance Manuals (ORISE 2002a, 2000a, and 2002b).
ESSAP surveys, their individual findings, and overall results are described in the
sections following.
(2) Surface Scans
Surface scans for beta and gamma radiation were performed over approximately
100 percent of the floor surfaces in the basement and on the first floor and 50
percent of the floor surfaces on the second floor. Surface scans for beta radiation
were performed over approximately 50 percent of the lower walls in the basement,
excluding the Stairwell General Area, 10 percent on the first floor, and 5 percent
on the second floor. Surface scans for beta radiation were also performed in the
vessel tunnel over approximately 50 percent of the surface.
Particular attention was given to remediated and adjacent surfaces, cracks and
joints in the floors and walls, and other locations where residual radioactive
material may have accumulated. Surface scans were not performed on any upper
wall or ceiling surfaces, in the Helium Rupture Disk Chamber, or in the Reactor
Building Ventilation Hold-Up Duct areas. Scans were performed using gas
proportional and NaI scintillation detectors coupled to ratemeters or ratemeter-
scalers with audible indicators. Locations of elevated direct radiation were noted
for further investigation.
ESSAP identified two areas of elevated beta surface radiation. One area was
found on a scabbled portion of the wall in the Bismuth Leak area. Another area
was found on the floor of the processor equipment room. The concrete block walls
in the air compressor room were also noted as being uniformly elevated. Scans of
the remaining surfaces did not identify any additional locations of elevated beta or
gamma radiation.
Surface scans of outdoor locations including soil areas, paved areas, and gravel
surfaces were performed over approximately 50 to 100 percent of the accessible
areas using a sodium iodide scintillation detector coupled to a ratemeter.
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Gamma surface scans were within the range of ambient background levels except
for an area adjacent to the NNRC that was determined to be caused by radiation
shine from the hot cell facility and storage vault.
(3) Surface Activity Measurements
Construction material-specific backgrounds were determined in areas of similar
construction, but without a history of radioactive material use. Ambient gamma
backgrounds were determined in areas where direct beta measurements were
performed; these background measurements were used to correct gross beta
surface activity measurements.
Direct measurements for total beta activity were performed at 35 locations, chosen
randomly and based on surface scan results. Additional measurements to
determine the average activity level in one area were also performed. Dry smears
were collected at each direct measurement location for determining removable
gross alpha and gross beta activity. Wet smears were collected from areas
adjacent to direct measurement locations to determine the H-3 and C-14 activity.
Direct measurements were performed using gas proportional detectors coupled to
ratemeter-scalers.
ESSAP identified an activity of 9,700 dpm/100 cm2 over approximately 0.5 m2 in
the elevated area identified in the Bismuth Leak area, with an average activity of
1700 dpm/100 cm2 over the contiguous one square meter area. The elevated
area identified in the process equipment room was limited to approximately
100 cm2 with an activity of 4,100 dpm/100 cm2. An activity range of 2,700 to
5,100 dpm/100 cm2 was determined for the concrete block in the air compressor
room, which GIT claimed resulted from naturally occurring radioactive material in
the blocks. Confirmatory scans on the interior and exterior of the room found the
radiation levels to be evenly distributed throughout the blocks, confirming the
activity was from the material used to make them. Removable activity levels
ranged from 0 to 3 dpm/100 cm2 for gross alpha and from -5 to 45 dpm/100 cm2
for gross beta. H-3 removable activity levels ranged from 3 to 466 dpm/100 cm2.
C-14 removable activity levels ranged from -2 to 86 dpm/100 cm2.
(4) Exposure Rate Measurements
ESSAP obtained background exposure rate measurements from various locations
within the NNRC, having similar construction as the GTRR. The NNRC has a site
history of radiological material usage; however, there are no other buildings similar
in construction to the GTRR and NNRC on the GIT campus. Exposure rate
measurements, using a microrem meter at one meter above the floor, were
performed in the center of selected areas or rooms within the GTRR.
Average interior building exposure rates ranged from 9 to 25 FR/h. Background
exposure rates performed in the NNRC ranged from 18 to 20 FR/h.
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Exterior exposure rate measurements, using a microrem meter at one meter
above the surface, were performed at five random locations from the reactor yard
area surrounding the GTRR.
Average exterior exposure rates ranged from 14 to 18 FR/h. Background
exposure rates performed at various intersections on the GIT campus ranged from
12 to 20 FR/h.
(5) Sampling
ESSAP collected surface soil (0-15 cm) samples at each exposure rate
measurement location.
Analysis of the soil samples by gamma spectroscopy for gamma-emitting mixed
fission and activation products identified Cs-137 at typical fall out concentrations.
Radionuclide concentrations for Co-60 and Cs-137, which are the predominant
radionuclides of concern at research reactor facilities ranged from -0.02 to 0.03
pCi/g for Co-60 and -0.02 to 0.21 pCi/g for Cs-137. All other radionuclides of
concern were reported as less than the respective minimum detectable
concentration of the procedure, which ranged from 0.03 to 0.11 pCi/g.
(6) ESSAP Results
Compliance for residual surface activity was shown using the GIT calibration
methodology approved by the NRC. Since ESSAPs calibration method differs,
this required adjusting the ESSAP-calculated surface activity by the ratio of the
efficiencies for the GIT and ESSAP methods. The correction factor was
approximately 2.3. All corrected ESSAP confirmatory surface activity
measurements, including the identified elevated areas, met guidelines and did not
require further remediation. Additional investigation by the inspector verified that
the concrete block in the air compressor room was made from material with a high
composition of naturally occurring radioactive material.
Except for the air compressor room in the basement, all exposure rate
measurements were less than 5 FR/h above background for each survey unit.
Confirmatory surface soil samples were less than the screening values listed in the
GIT final survey plan (GTS 2000).
c. Conclusions
Based on the above observations, surveys, evaluations, and analyses, the inspector
concluded that: 1) the elevated surface activity and exposure readings in the
basement compressor room were due to naturally occurring radioactive material; and
2) based on the results of the licensees final status survey and ESSAPs confirmatory
measurements, GIT has adequately demonstrated that the GTRR facility satisfies the
criteria for release for unrestricted use.
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8. MAINTENANCE AND SURVEILLANCE
a. Inspection Scope (IP 40755)
The inspector reviewed selected aspects of:
- maintenance procedures
- equipment maintenance records
- surveillance and calibration procedures
- surveillance, calibration, and test data sheets and records
- reactor periodic checks, tests, verification, and decommissioning activities
- facility design and DP changes and records
- NNRC Procedure 4200, 10 CFR 50.59 Review Program for Changes and Tests
During Decommissioning, Revision 01, dated November 1, 1999
- TS, Amendment No. 14, dated July 22, 1999
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b. Observations and Findings
(1) General Maintenance
During decommissioning general maintenance was focused on the support
services and equipment and not on any reactor systems. The inspector reviewed
maintenance records, interviewed staff and observed minor maintenance
performed on the various systems in operation. Based on the inspectors
interviews and observations, general maintenance was acceptable for an
industrial site.
(2) Surveillance
The inspector reviewed records of the TS Section 3 required surveillance
verifications performed during 2000. The results of the surveillances for the
radiation monitoring system and the ventilation system were within prescribed TS
limits and procedure parameters, and in close agreement with the previous
surveillance results.
(3) Change Control
TS or DP related 10 CFR 50.59 changes required review by the TSRC in
accordance with TS Section 5.2.
The inspector reviewed various TSRC approved change packages for changing
the method of accomplishing certain decommissioning activities. The inspector
determined that the changes had been evaluated, reviewed, and approved as
required by NNRC Procedure 4200, 10 CFR 50.59 Review Program for Changes
and Tests During Decommissioning, Revision 01, dated November 1, 1999. The
reviews were technically complete and adequately documented. Additionally, the
inspector concluded that TSRC 10 CFR 50.59 reviews and approvals were
focused on safety, and met licensee program requirements.
c. Conclusions
The licensee's program for surveillance and limiting conditions for operation
verification satisfied TS and DP requirements. The licensee's maintenance and
design change programs were in place and were being implemented as required by
licensee procedures.
9. RADIATION PROTECTION
a. Inspection Scope (IPs 69001 and 40755)
The inspector reviewed selected aspects of the radiation protection program (RPP):
- Radiation Protection Training
- radiological signs and posting
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- facility and equipment during tours
- routine surveys and monitoring
- survey and monitoring procedures
- dosimetry records
- maintenance and calibration of radiation monitoring equipment
- periodic checks, quality control, and test source certification records
- NNRC Radiation Protection Program (RPP)
- event/incident records
b. Observations and Findings
(1) Radiation Protection Program
Although individual procedures had been revised and some added, the RPP had
not functionally changed since the last inspection. The licensee reviewed the RPP
at least annually in accordance with 10 CFR 20.1101(c). This review and
oversight was provided by the TSRC as required by TS Section 5.2.d(9) and DP
Section 2.4.3.
The inspectors review of procedure change records, revisions, and radiation work
permits (RWP), confirmed that the RSO, individually and as a TSRC member,
reviewed and approved RWPs, and advised the Director and TSRC on matters
regarding radiological safety as required by TS Section 5.1.b, DP Section 2.4.1,
and the RPP.
Through record reviews and interviews with GTRR and Duratec staffs, the
inspector confirmed that the RPP was applied to all activities during the
decommissioning project, as required by DP Section 3.1 and GTRR procedures.
(2) Radiation Protection Postings
The inspector observed that caution signs, postings and controls to radiation and
contaminated areas at the NNRC were acceptable for the hazards involved and
were implemented as required by 10 CFR Part 20, Subpart J. The inspector
observed licensee and contractor personnel and verified that they complied with
the indicated precautions for access to such areas. The inspector confirmed that
current copies of NRC Form-3 and notices to workers were posted in appropriate
areas in the facility as required by 10 CFR Part 19.11.
(3) Radiation Protection Surveys
The inspector audited the GTRR daily, monthly, quarterly, and other periodic
contamination and radiation surveys, including airborne activity sampling,
performed from 2000 to 2003. The surveys were performed and documented as
required by DP Section 3.0, and GTRR survey procedures. HP surveys required
for special decommissioning activities, such as RWPs, were also performed and
documented as required. Results were evaluated and corrective actions taken and
documented when readings/results exceeded set action levels.
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-19-
(4) Dosimetry
The inspector confirmed that dosimetry was issued to staff, contractors, and
visitors as outlined in licensee procedures. The licensees dosimetry issuing
criteria specified that dosimetry should be issued to individuals who might receive
a dose equivalent exceeding 10 percent of the annual limits specified in 10 CFR
Part 20.1201(a). This criteria meet the requirements of 10 CFR 20.1502 for
individual monitoring. Training records showed that personnel were acceptably
trained in radiation protection practices. During the inspection the inspector
observed that workers and staff wore their dosimetry as required.
The licensee used a National Voluntary Laboratory Accreditation Program-
accredited vendor to process personnel thermoluminescent dosimetry. Dosimetry
results were reviewed by the RSO and doses above the facilitys ALARA limits
were investigated as required. The inspectors review of the licensees radiological
exposure records from 2000 to 2003 verified that occupational doses were within
10 CFR Part 20 limitations.
(5) Radiation Monitoring Equipment
The calibration and periodic checks of the portable survey meters, radiation
monitoring, air sampling, and counting lab instruments were performed by facility
staff or by certified contractors. The inspector confirmed that the licensees
calibration procedures and annual, quarterly, semiannual and monthly calibration,
test, and check frequencies satisfied TS Section 4.3.3, DP Section 3.1, and
10 CFR 20.1501(b) requirements, and the American National Standards Institute
N323 Radiation Protection Instrumentation Test and Calibration or the
instruments manufacturers' recommendations. The inspector verified that the
calibration and check sources used were traceable to the National Institute of
Standards and Technology and that the sources geometry and energies matched
those used in actual detection/analyses.
The inspector also reviewed Duratec instrument calibrations. Their calibration and
periodic checks of the portable survey meters, radiation monitoring, air sampling,
and counting lab instruments were performed by their staffs or by certified
contractors. The inspector confirmed that calibration procedures and annual,
semiannual quarterly, monthly, and daily calibrations, tests, and check frequencies
satisfied Duratec HPS procedures. Calibrations also met 10 CFR Part 20.1501(b)
requirements, and the American National Standards Institute N323 Radiation
Protection Instrumentation Test and Calibration or the instruments manufacturers'
recommendations. The inspector verified that the calibration and check sources
used were traceable to the National Institute of Standards and Technology and
that the sources geometry and energies matched those used in actual
detection/analyses.
The inspector reviewed the calibration lists and confirmed that calibrations for the
radiation monitoring and counting lab equipment in use had been performed and
that all portable instruments in use were calibrated.
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All instruments checked by the inspector had current calibrations appropriate for
the types and energies of radiation they were used to detect and/or measure.
(6) Respiratory Protection
DP Section 3.1.6 states that the Respiratory Protection Program will be
implemented by the decommissioning contractor in compliance with ANSI Z-88.2,
US NRC Regulatory Guide 8.15, 10 CFR 20.1701 through 20.1704, and OSHA
requirements.
While conducting inspections during decommissioning activities at the facility, the
inspector reviewed the respiratory protection program in use by contractor
personnel. The inspector noted that the licensee and contractor had established a
respiratory protection program as required by DP Section 3.1.6 and were using
tested and certified NIOSH/MSHA equipment as required. Records and
observation showed that air sampling was being conducted, surveys and
bioassays were completed as required, testing of respirators was being done, fit
testing of individuals was performed, and individuals were required to pass a
physical in order to qualify to use a respirator. The respiratory protection program
was in compliance with 10 CFR 20.1703 and the DP.
(4) Effluents
The program for the monitoring and storage of radioactive liquid, gases, and solids
was acceptable. Radioactive effluents were monitored and released when within
established limits as outlined in licensee procedures and the regulations. The
principles of As Low As Reasonably Achievable (ALARA) were acceptably
implemented to minimize radioactive releases. Monitoring equipment was
maintained and calibrated as required. Records were current and acceptably
maintained.
c. Conclusions
Based on the observations made and records audited, it was determined that,
because: 1) surveys were completed and documented as required by
10 CFR 20.1501(a) and licensee procedures, 2) postings met regulatory requirements,
3) the personnel dosimetry program was acceptably implemented and doses were in
conformance with licensee and 10 CFR Part 20 limits, 4) portable survey meters,
radiation monitoring, and counting lab instruments were maintained and calibrated as
required, 5) the evaluation and administration of the respiratory program were
adequately performed, and 6) the program for monitoring, storage, and release of
effluents was acceptable, the RPP implemented by the licensee satisfied NRC and DP
requirements.
5. EXIT MEETING SUMMARY
The inspector presented the inspection results to members of licensee management at
the conclusion of the inspection on October 23, 2002. The licensee acknowledged the
findings presented and did not identify as proprietary any of the material provided to or
reviewed by the inspector during the inspection.
PARTIAL LIST OF PERSONS CONTACTED
- T. Bauer Project Leader, ESSAP
- T. Brown Field Staff, ESSAP
- R. Eby Executive Engineer, (Vice President Energy, Environment, and Systems)
CH2M HILL
- N. Hertel Director, Neely Nuclear Research Center
- R. Ice Manager, Office of Radiation Safety
P. Jones Project Manager, GTS Duratek Field Services
G. Kalinauskas Senior Project Engineer, IT Corporation
R. Morton Field Staff, ESSAP
- Attended exit meeting.
The inspector also contacted other supervisory, technical and administrative staff personnel
as well.
INSPECTION PROCEDURE (IP) USED
IP 69001 Class II Non-Power Reactors
IP 40755 Class III Non-power Reactors
IP 85102 Material Control and Accounting - Reactors
IP 86740 Inspection of Transportation Activities
ITEMS OPENED AND CLOSED
Open
None
Closed
None
PARTIAL LIST OF ACRONYMS USED
Duratec GTS Duratec
DP Georgia Institute of Technology Research Reactor Decommissioning Plan dated
June 1998
ESSAP Environmental Survey and Site Assessment Program
GIT Georgia Institute of Technology
GTRR Georgia Institute of Technology Research Reactor
HP Health Physics
NNRC Neely Nuclear Research Center
NRC Nuclear Regulatory Commission
ORISE Oak Ridge Institute for Science and Education
RWP Radiation Work Permits
RPP Radiation Protection Program
RSO Radiation Safety Officer
TS Technical Specifications
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TSRC Technical Safety Review Committee